Roadmap for Indian activities

Abhishek Basak Bhabha Atomic Research Centre

India

Electricity in India

Between April, 2014 and January, 2015 the electrical energy availability in India was 868,591 MU against a requirement of 903,104 MU

Per capita consumption less than a third of world average

Shortages: Energy & Peak Power

Power cuts, load shedding

Huge demand & growing further

Current share of nuclear energy: 3.6%

Nuclear share must increase multifold to meet energy demands without leading to unmanageable environmental burden

2

Three Stage Nuclear Power Programme

STAGE 1 STAGE 2 STAGE 3

U - 233

ELECTRICITY

Depleted U

Pu

500 GWe

Pu FUELLED

FAST BREEDERS

Th

Very Large

ELECTRICITY

U - 233 FUELLED

REACTORS

Natural

Uranium ELECTRICITY

PHWR

10 GWe

Th

Pu

U - 233

3

Three Stage Indian Nuclear Power Programme – Current Status

Sta

ge-1

(C

urren

t G

en

erati

on

Th

erm

al

Reacto

rs)

Sta

ge-2

(Fast

Breed

er R

eacto

rs)

Sta

ge-3

(Th

oriu

m B

ased

Reacto

rs)

18 PHWRs operating

4 PHWRs under construction

3 LWRs under operation

1 LWR under commissioning

1 FBTR under operation since 1985

One 500 MWe PFBR in advanced stage of construction

KAMINI (30 kWt) operating since 1996

AHWR Critical Facility operating since 2008

AHWR construction to be launched

Current installed capacity: 5780 MWe

Plan to install 13,480 MWe by 2022 on progressive completion of projects under commissioning /construction and new projects that have been sanctioned

4

Additional reactors under pre-project stage

Project Location Capacity (MW)

Indigenous Reactors

CMAPP 1&2 Chutka, Madhya Pradesh 2 x 700

Mahi Banswara, 1&2 Mahi Banswara, Rajasthan 2 x 700

Kaiga 5&6 Kaiga, Karnataka 2 x 700

FBR 1&2 Kalpakkam, Tamilnadu 2 x 500

AHWR Location to be decided 300

Reactors with Foreign Cooperation

JNPP 1&2 Jaitapur, Maharashtra 2 x 1650

Kovvada 1&2 Kovvada, Andhra Pradesh 2 x 1500

Chhaya Mithi Virdi 1&2 Chhaya Mithi Virdi, Gujarat 2 x 1100

5

PHWR evolution

6

Performance of PHWRs

Operational

High Availability Factors- more than 90%

High Plant Load Factors- above 80%

10 reactors recorded continuous run of >1 year

RAPS5: 765 days of continuous operation (2nd longest continuous run in history, and longest in last two decades)

Safety

Over 396 reactor-years of safe operation

No incidence of radioactivity release beyond stipulated limit of the Regulatory Body AERB

Peer Reviews by WANO

IAEA OSART Team Review of RAPS 3&4

Positive report by IAEA Operation Safety Review

Team (OSART) team

7

Advanced Heavy Water Reactor (AHWR) fulfils safety requirements for the next generation nuclear energy systems

Salient features:

• Power level- 300 MWe • Main Heat Transport loop height- 39 m • Coolant- Boiling light water • Heat removal by natural circulation • Moderator: Heavy water • Lattice pitch: 225 mm square • Number of coolant channels: 452 • Small local peaking factors (flat fuel flux distribution) • Two independent functionally diverse shut down systems • Passive Poison Injection directly into fuel cluster as an additional

shutdown system • Small excess reactivity (On-power refuelling) • Number of control rods- 24

• Indian Regulatory body AERB completed pre-licensing safety appraisal of AHWR • AHWR-LEU design was peer reviewed by an expert committee • International recognition as an innovative design

Major design objectives: AHWR (Ref.)

A large fraction (~60%) of power from thorium

Extensive deployment of passive safety features – 7 days grace period, and no radiological impact in public domain

Addresses insider threat Design life of 100 years Easily replaceable coolant channels

Peak clad temperature hardly rises even in the extreme condition of complete station blackout and failure of primary and secondary shutdown systems

AHWR is a vertical pressure tube type, boiling light water cooled & heavy water moderated reactor using 233U-Th MOX & Pu-Th MOX in closed fuel cycle and AHWR-LEU using LEU-Th MOX fuel in once through fuel cycle

AHWR (Reference) AHWR-LEU

•Fuel cluster- 30 pins of (Th-Pu)O2 , 24 pins of (Th-233U)O2, & 2 pins of (Th-Pu)O2 doped with Gd (facilitate on-power refuelling) •Fuel burn up 40 GWd/te •Negative coolant void reactivity •Sufficient self-sustenance in

233U •Closed fuel cycle

•Fuel cluster- 54 pins of (Th-LEU)O2 & 2 pins of innermost ring are doped with Gd (facilitate on-power refuelling) •Fuel burn up: 60 GWd/te •All reactivity feedbacks are negative •Once through fuel cycle; as an export model •LEU contains 19.75% 235U •Discharged fuel is proliferation resistant (232U & 238Pu) •Very small amount of MA in the discharged fuel

0 10 20 30 40 50 600

10

20

30

40

50

60

70

80

90

Pow

er

fro

m thorium

(%

)

Burnup (GWd/t)

AHWR

AHWR-LEU

Role of AHWR passive systems to address insider

threat scenario

0 10 20 30 40 50 600

5

10

15

20

25

30

35

40

45

He

avy m

eta

l in

g/K

g

Fuel Burn-up GWd/Te

U-233 (AHWR)

Total Pu (AHWR)

U-233 (AHWR-LEU)

U-235 (AHWR-LEU)

Fuel [(Th, 233

U) MOX]

Fuel [(Th, Pu) MOX]

Fuel [(Th, Pu) MOX] with Gd

Displacer rod

Displacer tube

Fuel [(Th, LEU) MOX]

Fuel [(Th, LEU) MOX] with Gd

Displacer rod

Displacer tube

0 25 50 75 100

120

160

200

240

280

320

Te

mp

era

ture

(0C

)

Time (days)

Clad Surface Temperature

AHWR can withstand Fukushima type scenario for more than 100 days

8

Indian HTR and MSBR programme

Primary S/D

system

Control rods

Burnup

compensation

system

Shell

(Niobium alloy)

Gas gaps

Lower Plenum

Fuel Tube

Beryllia

Moderator

Graphite

Reflector

Upper

Plenum

Heat Pipe

CentralReflector

De-Fuelling Chute

Side Reflector

BottomReflector

Core Barrel

Support

Fuelling pipe

CoolantOutlet

Pebbles &Coolant

CoolantInlet

Reactor Vessel

Coolant

PebbleRetainingMesh

Secondary coolant salt inlet/ outlet

Blanket salt

Inlet/outlet

Reflectors

Primary HX

Core

To fuel salt reprocessing

Motor of primary circulation pump

Sacrificial salt layer

Reactor Vessel

Core

Blanket

Safety Vessel

Blanket Salt Circulation Pump

Blanket Salt Heat Exchanger

Blanket Salt Dump Line

Freeze Valves Fuel Salt

Dump Line

Intermediate Salt Dump

Line

Fuel Salt Intermediate

Heat Exchanger

Fuel Salt Circulation

Pump

Intermediate salt line

Fuel Salt

Blanket Salt Intermediate Salt

Inner Vessel

(a) (b) (c)

(d)

(a) Compact High Temperature Reactor, 100 kWt reactor for technology demonstration

(b) 600 MWth IHTR for large scale hydrogen production

(c) And (d) Pool Type and Loop-In-Tank type IMSBR concepts under study as options for Third Stage

9

Accelerator Driven sub-critical reactor Systems for Thorium Utilisation

Accelerator Driven Sub-critical reactor System (ADSS) is being developed for Thorium utilisation as well as for transmutation of nuclear waste in dedicated minor actinides burner reactor with inherent safety against power excursions. Development includes following three major elements

Development of a high current high energy proton accelerator

Stage-2: 1 GeV and 30 mA superconducting linac:

A major program for augmentation of the infrastructure required for development of a 1 GeV linac has also been launched

Stage-1: 30 mA 20 MeV Linac injector (LEHIPA)

Low Energy Beam Transmission line

Drift Tube Linac 1.3 MW Klystron

RFQ

60 kW RF System

Target development studies related to heat removal and window damage (irradiation creep and void swelling)

LBE thermal-hydraulic experimental test facility

Reactor program Basic theoretical studies,

advanced computer codes and compilation of nuclear data

Experimental program for sub critical ADS, including sub criticality measurement and monitoring

Several ADS concepts have been evolved: one way coupled, thorium burner, power producing Th breeder, Molten Salt Breeder ADS.

Experimental sub critical facility

10

Evolution of thorium fuel cycle development in India

KAMINIresearch reactor:

233U-Al Fuel

Dismantling of irradiated bundle

and Post-irradiation

examinationThoria pellets irradiation in

research reactor

Use of thoriain PHWR Irradiated fuel

reprocessing and fabrication

AHWR critical facility

Thorium extraction 11

Thorium extraction

Thorium Processing Monazite, which contains about 9% ThO2 and 0.35% U3O8, is a phosphate of thorium, uranium and rare earth elements. Thorium is extracted with trace uranium as by-product

Mining of thorium

Beach sands of India contain rich deposits of monazite (thorium ore), ilmenite, zircon, etc. The total minerals established so far include about 11.93 million tonnes of monazite containing 1.07 Million tonne of Thorium oxide.

Beach Sand Minerals

ILMENITE & RUTILE : TITANIUM (52%)

ZIRCON : ZIRCONIUM (3.2%)

MONAZITE : THORIUM &

RARE EARTHS (1.13%)

SILLIMANITE : ALUMINIUM

Composition of Monazite

Thorium as ThO2 9%

Rare Earths as REO 58.5%

Phosphorus as P2O5 27%

Uranium as U3O8 0.35%

Calcium as CaO 0.5%

Magnesium as MgO 0.1%

Iron as Fe2O3 0.2%

Lead as PbO 0.18%

Insolubles 3% 12

Thorium in Indian reactors

KAMINI: Research reactor operating at 30 kW power, commissioned at Kalpakkam in 1996. Reactor based on 233U fuel in the form of U-Al alloy, for neutron radiography.

Three PHWR stations at Kakrapar, Kaiga and Rajasthan (units 3&4) have irradiated a total of 232 thorium bundles, to maximum discharge burnup of 14 GWd/te. The power produced by the bundle just before discharge (600 FPD) was about 400 kW.

PURNIMA-II (1984 -1986): First research reactor using 233U fuel.

PURNIMA-III (1990-93): 233U-Al dispersion plate type fuel experiments.

CIRUS:

• Thoria rods irradiated in the reflector region for 233U production

• Irradiations of (Th, Pu) MOX fuels in Pressurised Water Loop to burnup of 18 GWd/te.

Thoria bundles irradiated in the blanket zone of Fast Breeder Test Reactor (FBTR)

233U-MOX fuel being irradiated in FBTR

13

-Autoradiograph , - Autoradiograph PIE hot cell facility Fission gas analysis set up

Post Irradiation Examination (PIE) of thoria fuel

The PIE was carried out for one of the discharged bundles from Kakrapar unit-2, which had seen 508 full power days.

• Dissolution tests done for uranium isotope composition and fission products and compared with theoretical evaluations.

• Power distribution shows peaking at the outer pins for UO2 bundle and at intermediate pins for thoria bundle.

• Fission products (137Cs) migrate to thoria pellet cracks unlike upto periphery in UO2 fuel.

0

100

200

300

400

500

0 5 10 15Burnup (GWd/te)

Bu

nd

le P

ow

er

(kW

)

Low Flux

Medium Flux

High Flux

UO2 bundle power 420 kW

Centre Intermidiate Outer

0

20

40

60

80

100

120

Rela

tive C

ounts

(C

s-1

37)

Pin position in bundle

UO2

ThO2

14

Fuel fabrication technology

Fuel Type Technique Type of facility

(Th-LEU) MOX Powder-Pellet Conventional UO2

(Th-Pu) MOX Powder-Pellet Glove Box

(Th-233U)MOX

Powder-Pellet Fully shielded facility

Sol-Gel Microsphere Pelletization Fully shielded facility

Pellet Impregnation Partially shielded facility

Coated Agglomerate Pelletization Partially shielded facility

Glove box facility for (Th-Pu) MOX

Sol-Gel facility for manufacturing microspheres Compaction press

Experience with fabrication of thoria-based fuel Thoria bundles for PHWRs Thoria assemblies for research reactor

irradiation Thoria based MOX pins for test irradiations Fabrication process was similar to that of UO2 & (U-Pu)MOX

Impregnation setup inside glove box

Attritor

Microspheres

15

Fuel reprocessing technology

Areas under study

Dissolution THOREX for industrial scale Third phase problems 3-component separation for

irradiated (Th,Pu)MOX Tackling gamma radiation from 232U

daughters products

Power Reactor Thorium Reprocessing Facility (PRTRF) Commissioned in January 2015 for reprocessing of thoria fuel bundles irradiated in PHWRs

Development of

THOREX Process

Flowsheet using

laboratory studies

Engineering scale facilities at Trombay for reprocessing irradiated Thoria fuel

Uranium Thorium Separation Facility (UTSF) Used for reprocessing thorium irradiated in research reactors

(a) Shielded glove boxes

(b) Laser Assisted Fuel Bundle Dismantling

(c) Shielded cubicle for processing of thorium lean waste

(d) Cell view of PRTRF

(a) (b) (c)

(d)

Actinide Separation

Demonstration Facility Industrial

scale facilities setup for

partitioning of high level liquid waste

16

AHWR critical facility for thorium based fuel physics studies

AHWR Critical Facility has been designed for conducting lattice physics experiments to validate AHWR physics calculations. There is enough flexibility to arrange the fuel inside the core in a precise geometry at the desired pitch for facilitating study of different core lattices based on various fuel types, moderator materials and reactivity control devices. This facility attained its criticality in April 2008.

The fuel cluster of present reference core consists of 19-pin metallic natural uranium fuel with aluminum cladding.

Features

• Thermal Neutron Flux (Ave) : 108 n/cm2/sec • Nominal Fission power : 100 Watts • Variable lattice pitch : ≥ 215 mm • Types of cores : Reference core, AHWR Core

and PHWR Core

Experiments performed Reference core • First approach to criticality; Dynamic tests for the shut down device • Measurement of critical height, level coefficient of reactivity etc. • Measurement of reaction rates and neutron spectrum • Radial flux profile in experimental cluster Thorium related experiments Experiments with (ThO2-U) Mixed Pin Cluster • Critical Height Measurement with one mixed Pin • Fine Structure Flux Measurement inside mixed pin cluster • Axial gamma scanning of the central ThO2 pin of mixed pin cluster • Level Coefficient of Reactivity Measurement with mixed pin cluster • Integral experiments with ThO2 -U mixed pin cluster and ThO2-PuO2

cluster Experiments with special Thorium based cluster • Critical height measurement experiments with one lattice location

fuelled with a six pin (Th-(1%)Pu) MOX cluster and a special six pin (Th-LEU) MOX cluster

• Axial flux measurements in (U-ThO2-U) Sandwich Cluster • Critical Height and Fine Structure Flux measurement experiments with

a specially designed (U-ThO2-U) Sandwich Cluster (In Extended Reference Core of AHWR-CF)

Cross-Sectional view of (ThO2-U) Mixed Pin Cluster

ThO2 pin

Nat. U pin

Critical facility-core

0 500 1000 1500 2000 2500

0

10

20

30

40

Ga

mm

a A

ctivity (

CP

S)

Elevation (mm)

Scan 1

Scan 2

Scan 3

17

KAMINI - the only reactor operating with 233U as fuel in the world now

KAMINI (KAlpakkam MINI) is a 30 kWth, 233U (20 wt%) -Al alloy fuelled, demineralised light water moderated and cooled, special purpose research reactor. Beryllium oxide (BeO) is used as reflector and cadmium is used as absorber material in the safety control plates. The reactor functions as a neutron source with a flux of 1012 n/cm2/s at the core center. Core cooling is by natural circulation of the coolant. It is used for:

•Neutron radiography of both radioactive and non-radioactive objects (e.g. FBTR fuel pins)

•Neutron activation analysis. •Carrying out radiation physics research, •Irradiation of large number of samples, and •Calibration and testing of neutron detectors.

Neutron radiography of irradiated FBTR control rod

Neutron radiography of space components

Chandrayaan (Indian lunar probe) mission critical devices were successfully inspected at KAMINI

Neutron radiography of MOX fuel pin

18

Indian collaborative efforts

As one of the seven Members of ITER

Components & Systems

TBM Development

CERN-LHC collaboration

Fermi Lab collaboration

Indo-French collaboration

Associated partner in Jules Horowitz Reactor

Foundation Course on Nuclear Energy for Bangladesh Atomic Energy Commission personnel

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Expected outcomes from ROADMAPS

Evaluate the expected timelines of advanced reactor systems in the global perspective

Identify potential stumbling blocks in the development and deployment of technologies, so that they can be addressed in development of similar technologies in India

Outline developments in Thorium utilisation technologies and non electric applications

Emphasise suitability of proven SMRs for use in niche areas

PHWR-220 has proven itself as a safe and reliable reactor with proven track record

Identify possible collaborative developmental areas

Identify reactor systems that have similar aims and characteristics.

It is entirely possible that aims and objectives of similar reactor systems may be entirely different – identify areas of overlapping technical collaboration.

20